Analysis of BEPU-Approached Multi-Physics Wall-to-Fluid Single Phase Friction Pressure Drop

Over the last few decades, the aim in Nuclear Reactor Safety (NRS) research and licensing has been to transition from conservative model use, which was widely used in the early stages of the nuclear industry, to the Best Estimate Plus Uncertainty (BEPU) approach in order to increase the accuracy of results to increase reactor output power safely. Furthermore, this decision is made to cover the increase in complexity of physics (i.e. Natural Circulation (NC), Passive Safety Systems) used in nuclear reactors, despite a reduction in the number of active mechanical parts. The central pillars of BEPU which are the selection of realistic Boundary and Initial Conditions (BIC) along with the adoption of an accurate model implemented in the Best Estimate (BE) System Thermal-hydraulic (SYS TH) Code, and uncertainty quantification methodologies should provide the reference for future works. Among the 116 Thermal-Hydraulic Phenomena (THP) for Water Cooled Nuclear Reactors (WCNRs) that are unique within the studied range of variables, the wall-to-fluid friction pressure drop (PD) as a Basic Phenomenon (BP) that can have a severe impact on the overall behavior of a thermal-hydraulic (TH) system. In fact, in normal operation conditions of NC systems, the efficiency of heat removal in the primary system is strongly dependent on PDs whereas, in off-normal scenarios, meeting Emergency Core Cooling System (ECCS) criteria may be subject to the influence of PDs along injection lines. Furthermore, this THP has an empirical origin, meaning there exists an integration domain in friction PD explicit formulas based on Re number and relative roughness, while some geometries used in Nuclear Power Proceedings of the 14th International Conference of the Croatian Nuclear Society Zadar, Croatia, June 9 – 12, 2024 Paper No. 120 SESSION X: Title of Session X (To be added by Programme Committee) Plants (NPP) may not fulfill the requirements of this integration domain. One of the main sources of uncertainty in friction factor calculations is the selection of an appropriate value for absolute roughness. The selection of the right value for roughness can become difficult because of the possible evolution of the surface of the hydraulic component concerning time. Indeed, the initial surface condition of the alloys can evolve on contact with the cooling solutions, by the growth of corrosion products, or by the deposition of particles resulting from this corrosion. Thus, roughness can change both with aging also based on different stages in subsequent time windows, and choosing a range for absolute roughness is suggested. This work's ultimate goal is to assess the possible errors in both RELAP Mod3.2mz and RELAP-3D as candidates for the most used BE SYS TH codes in Thermal-Hydraulic analysis and to emphasize the selection of the most appropriate model for wall-to-fluid friction factor. To do so, research focused on the results obtained in single-phase PDs using subcooled water with realistic BICs, since the correlations used for single-phase calculations are fairly well developed, and enough experimental data are available. In order to use the results obtained from codes in the nuclear industry, V&V&C must be applied in core-related safety issues, as it must be able to show the capability of computational tools with a demonstration of an error. As stated, wall-to-fluid friction PDs, must not be treated as low-level objectives anymore, especially for the design of RCS that rely on NC or passive systems. In addition, wall-to-fluid friction PD must be considered as a Multi-Physical BP, since both water chemistry and wall material, as representative of coolant chemistry and material, all affect the PDs.